This invention concerns a method for predicting the relative hydriding within a group of zirconium materials under nuclear irradiation. More specifically, it concerns a method for testing samples of zirconium materials outside of a nuclear reactor to determine their relative hydriding efficiencies under irradiation.
Zirconium metal alloys are widely used in core components and structures of water cooled nuclear fission reactors because of their low neutron cross section, among other apt properties for such service. Note for instance U.S. Pat. No. 4,212,686. Several zirconium alloy compositions have been developed and marketed primarily for nuclear reactor applications. Typical of such alloy compositions of zirconium are the commercially available materials identified as Zircaloy-2 and Zircaloy-4, comprising alloys set forth in U.S. Pat. Nos. 2,772,964 and 3,148,055. A niobium containing alloy of zirconium for reactor service is disclosed in U.S. Pat. Nos. 3,150,972 and 4,212,686.
The Zircaloys comprise alloy compositions containing at least about 95 percent by weight of zirconium metal and including in percent by weight up to about 2.0 percent of tin, up to about 0.5 percent of iron, up to about 0.5 percent of chromium and 0 to about 0.15 percent of nickel.
It is a problem that zirconium alloy materials absorb hydrogen generated by a corrosion reaction which occur under irradiation. The absorption of hydrogen causes embrittlement of the metal which is believed to be one of the most important factors limiting the life of zirconium alloys in pressurized water reactors. 5
Variations in the hydriding properties of light water reactor (LWR) cladding materials and other zirconium-based components are thus a concern in predicting the behavior of such materials in reactor service. Hydriding propensities sometimes vary substantially among materials that meet the specifications for a given alloy system. While lot-to-lot variations are not unique to zirconium-based systems, they have resulted in hydriding rates that vary by an order of magnitude for Zircaloy-2 materials in the same reactor environments [Johnson et al., Radiation Enhanced Oxidation of Zircaloy-2 in pH/10/LiOH and pH 10 NH.sub.4 OH, BNWL-463, Battelle Pacific Northwest Laboratory, Richland, Wash. (1967); Johnson, Zirconium Alloy Oxidation and Hydriding Under Irradiation: Review of Pacific Northwest Laboratory's Test Program Results, CPRI NP-5132, Electric Power Research Institute, Palo Alto, Calif. (1987), Lanning et al., "Waterside Corrosion Hydrogen Pickup, and Hydrogen Redistribution in Zircaloy-2 Pressure Tubes During Long Exposure in N Reactor", Third International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Aug. 30-Sept. 3, 1987, Traverse City, Mich.].
The fabrication and compositional variables that account for lot-to-lot differences are only partially understood. For example, elimination of nickel has been regarded as a key factor to improve Zircaloy hydriding resistance. However, Cheng et al. [International Symposium in Nuclear Power Systems - Water Reactors, p. 274, NACE/AIME/ANS, Myrtle Beach, S.C. (1984)] propose that fabrication plays a role in the hydriding property differences of Zircaloy-2 and Zircaloy-4.
Kass et al. demonstrated that increasing silicon content improves Zircaloy-2 resistance to out-of-reactor hydriding [Effects of Silicon, Nitrogen, and Oxygen on the Corrosion and Hydrogen Absorption Performance of Zircaloy-2, WAPD-283, Bettis Atomic Power Laboratory, Pittsburgh, Pa. (1963)], but provided no basis to judge the efficacy of silicon to suppress hydriding of Zircaloy-2 under irradiation.
U.S. Pat. No. 4,440,862 (Cheng et al.) describes an out-of-reactor procedure for testing zirconium alloys. But, that test concerns nodular corrosion and is in no way useful to predict hydriding characteristics [Johnson et al. (1967); Johnson (1987)].
Thus, there remains a need for an out-of-reactor test to predict zirconium alloy hydriding during reactor service. Such a test could be used in cladding fabrication to discriminate hydriding characteristics of various lots of currently used alloys and in research for optimization of alloys with respect to hydriding. Such optimization is needed since the trend to higher fuel burnups is resulting in cladding hydrogen contents above 500 ppm and associated loss of ductility [Pyecha et al., "Waterside Corrosion of PWR Fuel Rods through Burnups of 50,000 MWd/MTU", ANS Topical Meeting on LWR Fuel Performance, Apr. 21-24, 1985, Orlando, Fla.].